DISCLAIMER: This is a working document provided by the Centers for Disease Control and Prevention (CDC) 

technical support contractor, SC&A, for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its Working Groups 

or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, draft or working documents 

are not final NIOSH or ABRWH products or positions, unless specifically marked as such. This document prepared 

by SC&A represents its preliminary evaluation on technical issues.

NOTICE: This document has been reviewed to identify and redact any information that is protected by the 

Privacy Act 5 USC §552a and has been cleared for distribution.

MEMORANDUM

TO: Advisory Board on Radiation and Worker Health, Work Group on Carborundum 

Company 

FROM: Robert Anigstein 

DATE: November 27, 2018 

SUBJECT: Audit of NIOSH Assessment of External Doses from Plutonium Fuel Pellets

Background 

On August 10, 2018, NIOSH released a white paper (Guido 2018) in response to our review of 

the NIOSH MCNP1 analyses of the external exposures of Carborundum Company2 workers to 

uranium-plutonium pellets handled in a glovebox (Anigstein 2016, Appendix B). The report was 

supported by 44 data files that were posted on the DCAS restricted website. The present 

memorandum will begin with comments on the NIOSH white paper, followed by a review of the 

supporting data files. The comments on the white paper are keyed to sections of that report. 

1 MCNP is a generic term that can be applied to the MCNP family of codes that includes MCNPX and 

MCNP6.

2 The Carborundum Company will be referred to as “Carborundum” in the rest of this memo.

Review of NIOSH White Paper

Section 2.1: Pellet Geometry

Although Guido (2018) is correct in stating that maximizing the density of the pellet maximizes 

the amount of fuel in a single pellet, this does not lead to a more conservative (i.e., claimantfavorable) assessment. The dose rates assigned to workers are based on exposures to a batch of 

pellets that contains 100 g of plutonium. The analysis first calculates the dose rates from a single 

pellet, then multiplies that rate by the number of pellets in a 100-g plutonium batch. Thus, the 

more massive an individual pellet, the fewer pellets in the 100-g batch. However, a denser pellet 

results in more self-shielding of radiation emitted from the interior of the pellet, which 

potentially reduces the external dose rate. Guido’s statement regarding the maximized density 

incorrectly implies a conservatism in the analysis, whereas it actually has the opposite effect. 

Given the small size of an individual pellet, however, we do not anticipate that this would have a 

significant impact on the results.

 
Memo –External Doses from Plutonium Fuel Pellets 2 SC&A – November 27, 2018

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Section 3.2: Photon Source Calculations 

According to Guido (2018), “the photon intensities and energies emitted by the radionuclides 

were obtained from Be et al. (2004) and the International Commission on Radiological 

Protection (ICRP 2008).” We examined the NIOSH file MassFr_2_AtomFr.R1.U24Pu5y.xlsx to 

determine which radionuclides made the greatest contributions to the external photon doses from 

the glovebox worker scenario.3

3 We note that this file does not represent the final results, which are presented by Guido (2018, Tables 3–

6) and in the file Neutron_CarbidePellet dose calcs - revised 2018-07-31.xlsx. We assume that the relative 

contribution of individual radionuclides to the final doses are little changed in the final analysis.

 We spot-checked the MCNP files for the four radionuclides that 

were the largest contributors: 235U, 237U, 239Pu, and 241Am. According to comments entered into 

these files, the gamma ray energies and intensities of the first three of these nuclides were taken 

from Be et al., while the x-ray spectra were taken from ICRP Publication 107 (ICRP 2008). 

However, contrary to Guido’s statement, the photon spectrum of 241Am, cited in the NIOSH 

MCNP file GB_CRBRNDM.AM.Ph.AM241_.i as “Lund - x- and gamma,” was downloaded from 

Firestone and Ekström (2004). This is noteworthy since 241Am contributes about 81% of the 

ambient dose equivalent, H*(10), to the glovebox operator. 

According to a comment in the MCNP file cited above, the 241Am spectra were downloaded in 

2011; however, according to the website (Firestone and Ekström 2004), the literature cut-off date 

for 241Am was March 1, 1994. Any data published after that date were not utilized in compiling 

the photon spectrum for that radionuclide. Although the website remains available, the data are 

no longer updated. We confirmed that the photon spectrum was, indeed, taken from the cited 

source. The spectrum in the MCNP file was noticeably different from the spectrum based on 

more recent data from the Los Alamos National Laboratory (2012), which are in turn based on 

the Evaluated Nuclear Structure Data Files maintained by the Brookhaven National Laboratory, 

the universally accepted repository of nuclear data. Although we confirmed that the doses 

calculated using these more current data were not significantly different from those calculated by 

NIOSH, the dose assessments should be based on current science. We did not verify the photon 

spectra of radionuclides other than 241Am because the cited data sources are more recent than 

that for 241Am, and because these nuclides make much smaller contributions to the doses from 

this scenario.

Section 4.1: Fluence to Dose Conversion Coefficients 

We checked the H*(10) fluence to dose conversion coefficients entered in the MCNP file 

GB_CRBRNDM.AM.Ph.AM241_.i against the conversion coefficients for the ambient dose 

equivalent, H*(10), from photon fluence listed by ICRP (1996, Table A.21). According to a 

comment in the MCNP file, the coefficients were calculated by multiplying the conversion 

coefficients for air kerma per unit fluence, Ka/Φ, of monoenergetic photons listed by ICRP 

(1996, Table A.1), by the conversion coefficients for the ambient dose equivalent, H*(10)/Ka, 

from air kerma listed by ICRP (1996, Table A.21, col. 2). However, such a calculation is not 

needed because the H*(10) conversion coefficients from photon fluence are listed in Table A.21, 

col. 5. These values are slightly different from those calculated by NIOSH because, as noted in a 

footnote to Table A.21, they are derived from later values of air kerma per unit fluence, which 

 
Memo –External Doses from Plutonium Fuel Pellets 3 SC&A – November 27, 2018

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are listed in Table A.21, col. 4. MCNP calculations of dose rates using both sets of conversion 

coefficients show that the coefficients used by NIOSH result in a reduction of approximately 2% 

in the H*(10) doses from 241Am. 

Finding 1: NIOSH used H*(10) conversion coefficients from photon fluence, based on 

outdated data, that resulted in a reduction of approximately 2% in the H*(10) doses from 

241Am. 

Section 6.0: Dose Scenarios and Annual Dose Assignment 

We compared the total annual photon and neutron doses listed by Guido (2018, Table 6) with the 

SC&A results presented by Anigstein (2016, Table B-2). The results of the comparison are 

shown in Table 1. 

Table 1. Ambient Dose Equivalent Rates from External Exposure to (U,Pu)C Pellets 

(mrem/h)

Distance

SC&A NIOSH Differencea

Photons Neutrons Photons Neutrons Photons Neutrons

1 ft 6.318 0.202 9.507 0.313 50.5% 54.7%

1 m 0.624 0.026 1.209 0.038 93.7% 47.3% a NIOSH ÷ SC&A – 1 

The large differences between the SC&A and NIOSH results led us to examine the NIOSH 

MCNP analyses. We first noted that the NIOSH analyses were performed using MCNP6, version 

6.1, while the SC&A results were obtained with MCNPX, version 2.7.0. We had been informed 

by John Hendricks, one of the developers of both MCNP versions, that for the types of problems 

we were solving, MCNPX was adequate and had the advantage of being 3 times faster than 

MCNP6.1.4

4 John S. Hendricks, personal communication with Robert Anigstein, SC&A, Inc., November 2013.

Nevertheless, we tested this assumption by using MCNPX and the NIOSH input file 

GB_CRBRNDM.AM.Ph.AM241_.i to simulate the H*(10) photon doses. The tally results in the 

NIOSH output file GB_CRBRNDM.AM.Ph.AM241_.o matched those in our analysis within the 

statistical uncertainties. 

We also noted that the analysis reported by Anigstein (2016, Appendix B) employed point 

detector tallies (type 5 in MCNP parlance), while the NIOSH tallies were based on track lengths 

in a simulated dosimeter (type 4 tallies). Both methodologies are used in dose calculations and 

should yield similar, though not necessarily identical, results. We modified the NIOSH input file 

by adding point detectors located at the centers of the simulated dosimeters and repeated the 

simulation, simultaneously accumulating both types of tallies. Since the dose contribution from 

each event was thus calculated by both methods, the results should be quite similar. Instead, we 

found that the type 4 tally results at distances of 1 ft and 1 m were 32% and 90% higher, 

respectively, than the corresponding type 5 tallies. We performed numerous analyses, varying 

various input formats and parameters to determine the reason for this discrepancy. We ultimately 

discovered the problem to be the result of the source exponential directional bias employed in the 

NIOSH analyses. This bias is expressed by the following code in the MCNP input file: 

 
Memo –External Doses from Plutonium Fuel Pellets 4 SC&A – November 27, 2018

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vec=0 1 0 

dir=d75 

si75 -1 1 

sb75 -31 1 

This bias, intended to make the MCNP analyses execute faster, had not been properly 

implemented in either MCNPX nor in MCNP6.1. Our research uncovered a reference to this 

function by Goorley (2013). Under the heading “Selection of Low Priority Bugs in MCNP6,” 

Goorley listed the following: 

14) Incorrect source biasing with -31 function. Artf23313. 

This was the function entered in the NIOSH input file. Goorley’s document was issued April 23, 

2013, prior to the release of MCNP6.1 on May 8, 2013, and was “intended for distribution with 

MCNP6 production release.” This bug was fixed in MCNP version 6.2, which was released 

November 29, 2017. It was addressed by Werner et al. (2018, Table 2), as shown in the 

following excerpt: 

Table 2. Bug Fixes for MCNP Version 6.2 

Tracking Number Category Description

artf23313 Source Incorrect source biasing if using the '-31' special function on the SB card

We next removed the lines of code listed above from the MCNP input file and repeated the 

simulation of H*(10) doses from 241Am. The resulting doses calculated by the type 4 tallies—the 

simulated dosimeters used by NIOSH—at distances of 1 ft and 1 m were 34% and 37%, 

respectively, of the values calculated using the faulty source biasing. In other words, the source 

biasing used in this MCNP simulation resulted in an almost threefold increase in the doses.

Finding 2: NIOSH used incorrect source biasing in the MCNP dose analyses.

Further Review of NIOSH MCNP Photon Analyses

Exposure Geometry 

After the source bias was removed, the point detector (type 5) tallies at 1 ft and 1 m were 59% 

and 15% higher, respectively, than the corresponding type 4 tallies that were based on track 

lengths in simulated dosimeters. The cause was the exposure geometry, which is illustrated in 

Figure 1. As shown in the figure, the fuel pellet is only 1/8 inch above the floor of the glovebox. 

The simulated dosimeter, which measures 2 x 2 inches in the vertical plane and is 2 mm thick, is 

centered on the bottom surface of the glovebox. As shown by the line-of-sight drawn from the 

center of the fuel pellet to the dosimeter that just clears the edge of the glovebox floor, almost 

one-half of the dosimeter is effectively shielded from the source by the floor. In contract, the 

point detector, which is located at the center of the dosimeter, is in full view of the pellet, except 

for the intervening glass window. The dosimeter at a distance of 1 m (not shown in the figure), is 

more exposed to the source, since the sloping line of sight would intercept it further below its 
Memo –External Doses from Plutonium Fuel Pellets 5 SC&A – November 27, 2018

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center. This helps explain the large discrepancy between the two tallies at 1 ft and the smaller 

discrepancy at 1 m.

Figure 1. Glovebox Geometry Used in NIOSH MCNP Analysis (scales in cm)

We propose a change to the exposure geometry that would resolve this anomaly, as well as 

constituting a more realistic model of the gloveboxes used at Carborundum. Figure 2 shows 

some of the gloveboxes used in the plutonium research facility at Carborundum (Saulino et al. 

1962a). As shown in the figure, each glovebox has two ports in which the worker inserts his 

hands. The plutonium would most likely have been positioned at the level of the center of the 

ports, perhaps halfway in between them. By scaling the elevation of the centers of the ports in 

the figure to the 3-ft average height of the glovebox (Saulino et al. 1962b), we estimate that the 

ports were 24 cm above the floor. We therefore revised the MCNP input file to place the center 

of the fuel pellet 24 cm above the glovebox floor, and placed both the point detectors and the 

centers of the simulated dosimeters at the same elevation. The resulting analysis of H*(10) 

photon doses from 241Am showed that the results of the type 5 point detector tally at a distance of 

1 ft from the pellet agreed with the results of the type 4 simulated dosimeter tally within 1.5%. 

This constitutes reasonable agreement, given the finite size of the dosimeter and the fact that the 

combined relative statistical uncertainty was 0.5%. The two tallies at 1 m agreed within 0.6%, 

while the combined relative statistical uncertainty was 1.4%. We thus believe that either type 

tally produces valid results if the fuel pellets are elevated well above the floor of the glovebox. 

The H*(10) photon doses from 241Am based on type 4 tally results at distances of 1 ft and 1 m at 

the level of the glovebox floor are 45% and 21%, respectively, lower than the doses calculated at 

an elevation of 24 cm above the floor. 

Finding 3. The simulated dosimeters in the glovebox geometry modeled by NIOSH are 

partially shielded by the floor of the glovebox, which reduces the calculated doses. 
Memo –External Doses from Plutonium Fuel Pellets 6 SC&A – November 27, 2018

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Figure 2. Portion of Work Area of Plutonium Research Facility Showing Glovebox 

Arrangement (Saulino et al. 1962a) 

Minor Technical Errors

We also observed two minor errors in the MCNP input file. The vertical extent of the source is 

specified as 0.5 cm, whereas the height of the pellet is specified as 0.51 cm. Thus, the upper 

0.1 cm of the cylindrical pellet is not sampled by MCNP. Although this is too small a 

discrepancy to affect the results, it is nevertheless worth mentioning in the interest of a 

technically correct analysis. Furthermore, the radial extent of the source is specified as 0.26 cm, 

while the radius of the pellet is given as 0.255 cm. The result is that MCNP attempts to sample a 

region beyond the actual pellet but, finding that this is outside the designated cell, rejects it, 

leading to a slight inefficiency in the execution of the program. Again, this has no impact on the 

results.

Review of NIOSH MCNP Neutron Analysis 

We also performed a review of the NIOSH MCNP neutron analysis. Examining the input file 

GB_CRBRNDM.AM.Ns.Carbide.i, we found the same erroneous source exponential directional 

bias function that was used in the photon simulations. We also found the same problem with the 
Memo –External Doses from Plutonium Fuel Pellets 7 SC&A – November 27, 2018

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glovebox exposure geometry discussed previously in the present memo. We revised the MCNP 

neutron analysis, eliminating the erroneous source bias function and elevating the pellet and the 

dose points 24 cm above the floor of the glovebox, as we did for the photon analysis. We found 

that the NIOSH neutron analysis shown in the MCNP output file 

GB_CRBRNDM.AM.Ns.Carbide.o produced H*(10) doses at distances of 1 ft and 1 m that were 

110% and 164% higher, respectively, than the type 4 simulated dosimeter tallies in the revised 

analysis. 

Conclusions

It is not feasible for us to directly evaluate the photon and neutron doses to workers that would 

result from the corrections and revisions discussed in the present memo. First, this would require 

us to revise the analyses of photon doses from 15 radionuclides in addition to 241Am. Second, as 

noted in footnote 3 of this memo, we do not have the final spreadsheets which NIOSH used to 

convert the MCNP results, which are expressed as pSv per nuclear disintegration, to dose rates 

from a batch of pellets in the glovebox. However, since we have evaluated the changes in the 

241Am photon dose rates, and since 241Am contributes over 80% of the photon dose, we can 

estimate the approximate changes to the photon doses by scaling the doses reported by Guido 

(2018) to the changes in 241Am H*(10) photon dose rates, according to the following formula: 

(1)

N

SN r d

dD D =

where

Dr = recalculated H*(10) photon dose rate (mrem/h)

DN = photon dose rate based on total over all energy ranges reported by Guido (2018, 

Table 6), assuming exposure for 1,000 h/y 

ds = dose rate from 241Am based on revised MCNP analysis

dN = dose rate from 241Am in GB_CRBRNDM.AM.Ph.AM241_.o 

The results of the calculation are shown in Table 3. We observe excellent agreement between the 

dose rates at 1 ft based on the revised MCNP calculations, as discussed in the present memo, and 

the dose rates originally calculated by SC&A. This gives us reason to believe that NIOSH is

correctly translating the MCNP results into hourly doses. The recalculated doses at 1 m, based on 

the NIOSH analysis, are 4% lower than the original SC&A analysis. 
Memo –External Doses from Plutonium Fuel Pellets 8 SC&A – November 27, 2018

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Table 3. Recalculated H*(10) Photon Dose Rates 

Distance

NIOSHa 

(mrem/h)

241Am (pSv per dis)b

 Recalculatedc

(mrem/h)

SC&Ad

(mrem/h)

Differencee

NIOSHf SC&Ag

1 ft 9.507 6.641e-07 4.412e-07 6.316 6.318 -0.03%

1 m 1.209 8.480e-08 4.197e-08 0.598 0.624 -4.10%

 a Based on Guido (2018, Table 6)

b pSv per nuclear disintegration

c Calculated using Eq. (1) 

d Anigstein (2016, Table B-2) 

e

 Col. 5 ÷ Col. 6 - 1 

f From GB_CRBRNDM.AM.Ph.AM241_.o

g Revised MCNP results for type 4 tallies
Memo –External Doses from Plutonium Fuel Pellets 9 SC&A – November 27, 2018

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